05000324/LER-2008-003, Regarding Reactor Building Crane Design Inadequacy

Regarding Reactor Building Crane Design Inadequacy

ML082320815
Person / Time
Site:	Brunswick
Issue date:	08/11/2008
From:	Wills E Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 08-0103 LER 08-003-00
Download: ML082320815 (6)

LER-2008-003, Regarding Reactor Building Crane Design Inadequacy

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3242008003R00 - NRC Website

text

Progress Energy AUG 1 12008 SERIAL: BSEP 08-0103 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Licensee Event Report 1-2008-003 Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Carolina Power

& Light Company, now doing business as Progress'Energy Carolinas, Inc., submits the enclosed Licensee Event Report (LER). This report fulfills the requirement for a written report within sixty (60) days of a reportable occurrence.

Please refer any questions regarding this submittal to Mr. Philip A. Leich, Manager -

Support Services, at (910) 457-2271.

Sincerely, Edward L. Wills, Jr.

Plant General Manager Brunswick Steam Electric Plant MAT/mat

Enclosure:

Licensee Event Report Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant PO Box 10429 Southport, NC 28461 A~J4I<

Document Control Desk BSEP 08-0103 / Page 2 cc (with enclosure):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Joseph D. Austin, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)

, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Brunswick Steam Electric Plant (BSEP), Unit 1 05000325 1 OF 4

4. TITLE Reactor Building Crane Design Inadequacy

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED I

FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.

MONTH DAY YEAR 05000324 SEUMENILR EVO BSEP, Unit 20002 FACILITY NAME DOCKET NUMBER 06 11 2008 2008 - 003 - 00 08 11 2008 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

F1 20.2201(b)

L1 20.2203(a)(3)(i)

[L 50.73(a)(2)(i)(C)

[I 50.73(a)(2)(vii)

[ _1 20.2201(d)

El 20.2203(a)(3)(ii)

[I 50.73(a)(2)(ii)(A) 11 50.73(a)(2)(viii)(A)

Ej 20.2203(a)(1)

El 20.2203(a)(4) 0 50.73(a)(2)(ii)(B)

El 50.73(a)(2)(viii)(B)

[__ 20.2203(a)(2)(i)

L-50.36(c)(1)(i)(A)

[I 50.73(a)(2)(iii)

El 50.73(a)(2)(ix)(A)

10. POWER LEVEL

[] 20.2203(a)(2)(ii)

El 50.36(c)(1)(ii)(A)

El 50.73(a)(2)(iv)(A)

[] 50.73(a)(2)(x)

[E 20.2203(a)(2)(iii)

El 50.36(c)(2)

El 50.73(a)(2)(v)(A)

El 73.71 (a)(4) 100 [E 20.2203(a)(2)(iv)

El 50.46(a)(3)(ii)

El 50.73(a)(2)(v)(B)

El 73.71 (a)(5)

El 20.2203(a)(2)(v)

E] 50.73(a)(2)(i)(A)

E] 50.73(a)(2)(v)(C)

El OTHER LI 20.2203(a)(2)(vi)

EL 50.73(a)(2)(i)(B)

[:1 50.73(a)(2)(v)(D)

Specify in Abstract below or I_

in

Event Cause

The direct cause of this event is that the crane girder end connection design was not adequately evaluated during the initial design of the crane by Whiting Corporation. The crane design was performed by Whiting Corporation in the 1970's. Due to the historical nature of this condition, determining a plausible cause is not practical or feasible.

Safety Assessment

The safety significance of this condition is minimal. This concern is only applicable to conditions present during an extremely unlikely DBE. However, the ability of the crane to maintain a load or structural stability during the DBE is in question. If a DBE had occurred prior to the implementation of the compensatory measures, there was a potential that the lack of structural integrity could have resulted in crane structural damage which, in turn, could have adversely impacted structures, systems, or components in the vicinity of the crane at the time. Due to the difficulty in predicting the various effects on plant equipment, this condition was conservatively considered an unanalyzed condition that significantly degraded plant safety.

Corrective Actions

Modifications have been implemented for the Unit 1 and Unit 2 Reactor Building cranes to allow continued restricted use of the cranes for loads of up to 40 tons.

Engineering Changes will.be developed and implemented to restore the cranes to their original seismic design requirements. These modifications are currently scheduled to be completed by February 27, 2009, for Unit 2 and July 31, 2009, for Unit 1.

Previous -Similar Events A review of LERs and corrective action program condition reports for the past three years identified the following similar event.

Nuclear Condition Report (NCR) 251648, originated on October 27, 2007, identified a similar concern related to the tornado wind loading design of the cranes. At the time this condition was i dentif ied, there was no indication that the original seismic design was in question. Additionally, in 1997, calculation OSEIS-0036 was performed. This calculation included a re-analysis of the Reactor Building crane structure. BSEP Engineering relied on calculation OSEIS-0036 to provide assurance that there were no seismic concerns with the cranes. However, as with the original evaluation, this analysis did not consider the effect on the girder end connectors. As a result the corrective actions for NCR 251648 focused on the wind loading issue and would not have been reasonably expected to prevent the condition discussed in this LER.

Commitments

No regulatory commitments are contained in this report.